The present invention relates to a method for immobilizing radioactive noble gases in a zeolite matrix wherein the noble gas, after heat treatment of the zeolite matrix, is forced under high pressure into the cavities of the structure of the zeolite matrix and is encapsulated in these cavities by cooling the matrix while maintaining the pressure.
The waste gases developed during reprocessing of irradiated nuclear reactor fuel elements contain certain quantities of radioactive noble gases, originating particularly from removal of the cladding of the fuel elements and subsequent dissolution of the fuel material. In the prior art, these noble gases, if they were separated from the waste gases at all, were introduced into pressurized steel bottles for transport to a location which permitted limited time storage.
The radioactive gas in such a pressurized bottle is under a high pressure, e.g. more than 100 bar, where 1 bar=0987 std. atm., spontaneously generates heat, and, depending on the radioactivity inventory in the bottle or the cooling mode, e.g., natural air convection, attains an increased temperature, e.g. 393.degree. K. Thus the wall of such a bottle is continuously subjected to substantial thermally induced tensile stress. If the cooling system were to malfunction or become inoperative, it is possible that the bottle would crack or burst, resulting in the release of the entire radioactive noble gas inventory being stored or transported.
The noble fission gas consists mainly of krypton and xenon isotopes. The daughter nuclide of krypton is rubidium, an alkali metal which is highly reactive and capable of inflicting corrosion damage. Rubidium and certain impurities possibly present in the noble gas, such as, e.g. oxygen, water, etc., react together and form products such as, for example, Rb.sub.2 O, RbOH, etc., which are even more corrosive than the alkali metal itself (the latter will be in a molten state at the storage temperatures, excepted during the first decades.
The grave drawbacks of storing radioactive noble gases in pressurized gas bottles on one hand and the large quantities of krypton-85 expected to have to be stored safely for long periods of time in the future, on the other, have made it necessary to search for alternative ways to carry out the long-term storage of highly radioactive noble gases. One alternative that has been proposed previously consists in the solidification of noble gases in zeolites or molecular sieves.
Zeolites or molecular sieves, have been used, for example, in the separation of mixtures of substances by means of gas chromatography, involving a large number of repeated alternations of adsorption and desorption processes. However, in the solidification of radioactive noble gases in zeolites, desorption must be avoided as much as possible because increased safety during transport and storage can be assured only if gas diffusion out of the loaded zeolite is only very slight. Essentially, the gas diffusion is determined by the type of zeolite, and by the temperature.
The temperature in the zeolite structure itself depends on the radioactive gas load in the zeolite and the heat transfer through the inorganic matrix/gaseous phase. A large number of tests have been made directed toward the selection of suitable zeolites and the best process conditions. Normally, molecules having a larger diameter than the channels or pores in a given zeolite are not sorbed by that zeolite. However, it has been found that by inceasing the temperature from room temperature to, for example, 770.degree. K., the pores of certain zeolites, as, for example, zeolite 3A or sodalite, are widened and krypton can be forced into these cavities in the crystal structure under a very high pressure, e.g. 2000 bar. If thereafter the system is cooled while maintaining the high pressure, the gas is encapsulated in the cavities. The encapsulated gas is then, in contradistinction to the condition existing in the case of adsorption, not in equilibrium with the gaseous phase.
A series of differently produced, leached and unleached sodalite types have been examined as to their capability to encapsulate krypton or krypton-xenon mixtures, they have been described in the report by R. W. Benedict, A. B. Christensen, J. A. Del Debbio, J. H. Keller, and D. A. Knecht: Technical and Economic Feasibility of Zeolite Encapsulation for Krypton-85 Storage; DOE Report No. ENICO-1011, September, 1979. In their encapsulation experiments, the authors employed temperatures between 670.degree. and 850.degree. K., and pressures between 1200 and 2000 bar.
In order to evaluate which zeolites were best suited for the encapsulation of krypton, untreated K-exchanged, Cs-exchanged, and Rb-exchanged zeolites A and various sodalite types were examined with respect to maximum loading as well as temperature and radiation resistance to gas diffusion out of the loaded zeolites (krypton leakages). Krypton loadings from 20 to 40 cm.sup.3 STP/g sodalite or zeolite A were found. The loading values for leached sodalite were higher than for unleached sodalite. Krypton leakage measurements were made over short times, i.e. about 2 to 24 hours, at temperatures between 570.degree. and 775.degree. K. and for longer periods, i.e. about 1 to 12 months, at a temperature of 423.degree. K. The lowest leakage rates were found:
(a) for samples with low adsorbed H.sub.2 O content compared to samples with high adsorbed H.sub.2 O content; PA1 (b) for samples with high initial krypton loading compared to samples with low loading; PA1 (c) for unleached sodalite compared to leached sodalite. PA1 (a) bringing the noble gas to be immobilized into contact with the zeolite in a vessel which was previously evacuated to a pressure of less than 1 mbar; PA1 (b) then forcing, or pressing, the noble gas into the cavities of the zeolite at a temperature in the range from 720.degree. K. to 870.degree. K. and under a pressure of 200 bar to about 2000 bar; and PA1 (c) finally, cooling the zeolite loaded with the noble gas in a known manner. PA1 (a) a reduction of gas diffusion from the loaded zeolite matrix so that even with high loads no pore closing methods (resin, glass etc.) need be employed and simultaneously safety is increased not only during the transport of the solidified radioactive gas but also during long-term storage; PA1 (b) sorption of in the noble gas at relatively low pressure, e.g., less than 600 bar; PA1 (c) fixing without compressors employing for instace a combination cryoautoclave/high pressure autoclave achieving consequently a reduction of potential leakage sources and a reduction of the free inventory of radioactive noble gas; and PA1 (d) recovery of the noble fission gas not fixed after pressing, remaining in the lines of the apparatus and the autoclav, by cryopump effect.
From the test results, Benedict et al drew the conclusion that for unleached sodalite with a krypton loading of about 20 cm.sup.3 /g and low quantities of adsorbed water, the predicted 10-year leakage of krypton at a final storage temperature of 423.degree. K. will be less than 0.1%.
Under consideration of the Kr-85 decay heat, sodalite (of the formula Na.sub.2 O.times.Al.sub.2 O.sub.3 .times.2SiO.sub.2 .times.2.5 H.sub.2 O) seemed to be sufficiently thermally stable after loading with noble gas to assure the immobilization of krypton-85 for more than 100 years without the use of a technically very complicated closing of pores, which could possibly be effected in the rolling or fluidized bed process with a still to be found radiation resistant resin. The long term thermal stability at temperatures above 423.degree. K. which initially had been determined theoretically by extrapolation on the basis of the activation energy for the gas diffusion out of the zeolite could, however, not be confirmed experimentally.
Tests with sodalite samples loaded with argon, (the effective kinetic diameters of Kr and Ar are very similar, i.e. 0.39 mm for krypton and 0.37 mm for argon), have shown that already at 473.degree. K. the stability towards elevated temperatures of the loaded sodalite samples is insufficient. Sodalite loaded with 30.5 cm.sup.3 STP Ar/g, looses 52% of the encapsulated gas at 473.degree. K. already after 1080 hours. This desorption, which is undesirable for final storage, can be counteracted only by restriction of loading or use of a pore closing resin. Lower loading, however, is associated with increased costs and increased waste volume. Furthermore, the homogeneous embedding of highly radioactive extrudates in a resin is a technically difficult undertaking.
Additionally, the recommended loading conditions, for example, a temperature of 773.degree. K. and a pressure of 2000 bars, are undesirable when working with large inventories of radioactive gases. Since the use of at least one compressor is required, the expenditures required to keep down leakages at the apparatus are considerable. A high pressure system which is complicated from a safety point of view becomes a prerequisite.